Boiling Water Reactors (BWR) designed for power generation utilize fuel assemblies arranged inside vertical channels through which water coolant is injected. Each of the fuel assemblies consist of a plurality of vertical rods arrayed within the said vertical channels. The said rods, henceforth called "fuel rods" are sealed cylindrical tubes inside which ceramic pellets of fissionable material, e.g. uranium oxide, are stacked. The fuel pellet stack rests on the bottom of the fuel rod tube and held down on the top by a compressed spring. The space above the fuel pellet stack where the spring is placed is called the fission gas plenum. The fuel rods are initially filled,with an inert gas at relatively low pressure, and the pressure inside the rod increases as the fission-produced gasses accumulate throughout the life time of the fuel.
The fuel rods are arrayed in a lattice of fixed dimensions. The position of each rod is held in the lattice by an upper and lower tie plates. To prevent fuel rod bowing and to hold the fuel rods in the lattice positions at all elevations, several so-called spacers are placed at nearly equally-spaced intervals along the length of the assembly. The water flows upward in the channels and removes the heat generated in the pellets by the fissioning of the heavy isotopes. In addition to its cooling function, the water serves as neutron moderator.
In boiling water reactors, the water is allowed to boil as it travels up in the fuel assembly channel. The density of water is reduced by the boiling process which results in adverse effects on both the neutronic and the thermal-hydraulic performance of the fuel assembly. The coolant boiling adverse impact on the neutronic performance is a result of the reduced neutron moderation efficiency particularly in the upper portion of the fuel assembly, where the fuel-to-moderator ratio becomes higher than optimally desired. This problem is usually mitigated by introducing one or more water rods or channels through which water flows without boiling and acts to improve neutron moderation.
The thermal-hydraulic performance is affected by the boiling of the coolant because the pressure drop for a flowing mixture of steam and water is significantly higher than the pressure drop for the same mass flow rate of liquid water. The ratio of the pressure drop of flowing steam/water mixture to the pressure drop of the same flow rate of liquid water is the well known two-phase pressure loss multiplier, henceforth called "2-.phi. multiplier." It is common to have 2-.phi. multiplier of 5.about.10 signifying large pressure drop due to boiling in the fuel assembly channel.
Modern designs of BWR fuel assemblies utilize a larger number of fuel rods compared with the early designs; a trend motivated by a variety of neutron economy and operational flexibility issues the discussion of which is beyond the scope of this document and is well-known to the practitioners of the art. The early fuel assembly designs consists of fuel rods arrayed in a 7.times.7 lattice. Subsequent designs utilize 8.times.8 and 9.times.9 lattices. Present designs use 10.times.10 fuel rod arrays and further increase in the number of fuel rods is anticipated. The increase in the number of fuel rods necessitates a decrease in the diameter of each rod and a decrease in the distance between rods in the lattice (known as rod pitch). As a result of the decrease in rod diameters and pitch, the dimensions of the so-called subchannels are similarly decreased. A subchannel is the cross sectional area of the flow domain surrounded by four rods, or rods and cold wall surface. The decrease of the subchannel dimensions can be equivalently expressed as a decrease in the so-called hydraulic equivalent diameter, henceforth called "hydraulic diameter." The hydraulic diameter is defined as four times the total flow area divided by the wetted perimeter. While fuel designers kept the change in flow area relatively small when the number of fuel rods is increased, the wetted perimeter is increased significantly resulting in a net decrease in the hydraulic diameter. It is well-known that the pressure drop for flow through a fuel assembly increases with the decrease of the hydraulic diameter. This increase in the 2-.phi. pressure drop for modern designs of fuel assemblies poses undesirable limitations on the maximum flow through a fuel assembly given a fixed pumping power.
The increase in 2-.phi. pressure drop of a fuel assembly design poses serious limitations on the stable operation of a BWR containing such fuel assemblies. It is well-known that the margin to instability is reduced when the ratio between the two-phase pressure drop to single-phase pressure drop increases.
The boiling in the BWR fuel assembly is a fundamental aspect of the operation of the power plant, and cannot in itself be reduced to produce favorable outcome. The reduction of the pressure drop resulting from such boiling is a much needed improvement. Such pressure drop reduction was achieved in a limited way in the prior art by using alternate designs and materials for the spacers in an effort to reduce their flow blocking effect. The spacer design impact on the total 2-.phi. pressure drop is limited because the 2-.phi. pressure drop is dominated by the friction in the bare fuel rod region.
The present invention introduces a totally different method and apparatus to achieve the desirable reduction of the 2-.phi. pressure drop by introducing design changes in the fuel rods themselves as explained in the following sections.